Quenching degradation in-pile experiment on an oxidized fuel rod in the temperature range of 1000 to 1260°C

Abstract

Under core uncovery accident conditions, the cladding tube of a fuel rod will be oxidized and embrittled. The fuel degradation conditions due to the thermal shock during delayed reflooding need to be studied. In the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute, the sequences in a severe accident were simulated to investigate the in-core fuel degradation due to quenching. With these in-pile experiments, the oxidation behavior of the Zircaloy cladding tube was clarified at temperatures ranging 1000–1260°C, and it was shown that there was fuel degradation due to the thermal shock by the reflooding after the cladding was exposed to high-temperature steam for a relatively long time. Analysis of the test results was also performed using the SCDAP code to evaluate the applicability of this code to these particular tests and to obtain supporting data for the test results. Generally, the calculated results agreed well with the test results. However, at lower elevation of the fuel rod, the predicted cladding temperature and oxide layer thickness overestimated the test results due to the modeling of the cooling effect by steam flow.